Over recent years, a substantial amount of interest has grown in developing commercially viable PWRs that utilize the phenomenon of natural circulation (also known as thermosiphon effect) to circulate the primary coolant to both cool the nuclear reactor and to vaporize a secondary coolant into motive vapor.
CAREM (Argentina) is a 100 MW(e) PWR reactor design with an integrated self-pressurized primary system through which the primary coolant circulation is achieved by natural circulation. The CAREM design incorporates several passive safety systems. The entire primary system including the core, steam generators, primary coolant and steam dome are contained inside a single pressure vessel. The strong negative temperature coefficient of reactivity enhances the self-controlling features. The reactor is practically self-controlled and need for control rod movement is minimized. In order to keep a strong negative temperature coefficient of reactivity during the whole operational cycle, it is not necessary to utilize soluble boron for burn-up compensation. Reactivity compensation for burn-up is obtained with burnable poisons, i.e. gadolinium oxide dispersed in the uranium di-oxide fuel. Primary coolant enters the core from the lower plenum. After being heated the primary coolant exits the core and flows up through the riser to the upper dome. In the upper part, the primary coolant leaves the riser through lateral windows to the external region, then flows down through modular steam generators, decreasing its enthalpy by giving up heat to the secondary coolant in the steam generator. Finally, the primary coolant exits the internal steam generators and flows down through the down-corner to the lower plenum, closing the circuit. CAREM uses once-through straight tube steam generators. Twelve steam generators are arranged in an annular array inside the pressure vessel above the core. The primary coolant flows through the inside of the tubes, and the secondary coolant flows across the outside of the tubes. A shell and two tube plates form the barrier between primary and secondary coolant flow circuits.
AST-500 (Russia) is a 500 MW(th) reactor design intended to generate low temperature heat for district heating and hot water supply to cities. AST-500 is a pressurized water reactor with integral layout of the primary components and natural circulation of the primary coolant. Features of the AST-500 reactor include natural circulation of the primary coolant under reduced working parameters and specific features of the integral reactor, such as a built-in steam-gas pressurizer, in-reactor heat exchangers for emergency heat removal, and an external guard vessel.
V-500 SKDI *(Russia) is a 500 MW(e) light water integral reactor design with natural circulation of the primary coolant in a vessel with a diameter less than 5 m. The reactor core and the steam generators are contained within the steel pressure vessel (i.e., the reactor pressure vessel). The core has 121 shroudless fuel assemblies having 18 control rod clusters. Thirty six fuel assemblies have burnable poison rods. The hot primary coolant moves from the core through the riser and upper shroud windows into the steam generators located in the downcomer. The coolant flows due to the difference in coolant densities in the downcomer and riser. The pressurizer is connected by two pipelines, to the reactor pressure vessel and the water clean up system.
The NHR-200 (China) is a design for providing heat for district heating, industrial processes and seawater desalination. The reactor power is 200 MW(th). The reactor core is located at the bottom of the reactor pressure vessel (RPV). The system pressure is maintained by N2 and steam. The reactor vessel is cylindrical. The RPV is 4.8 m in diameter, 14 m in height, and 197 tons in weight. The guard vessel consists of a cylindrical portion with a diameter of 5 m and an upper cone portion with maximum 7 m in diameter. The guard vessel is 15.1 m in height and 233 tons in weight. The core is cooled by natural circulation in the range front full power operation to residual heat removal. There is a long riser on the core outlet to enhance the natural circulation capacity. The height of the riser is about 6 m. Even in case of interruption of natural circulation in the primary circuit due to a LOCA the residual heat of the core can be transmitted by steam condensed at the uncovered tube surface of the primary heat exchanger.
While the aforementioned PWRs utilize natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant, all of these natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service but also subjects the equipment to corrosive conditions. Furthermore, locating the heat exchange equipment within the reactor pressure vessel results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. However, prior to the present invention, the location of the heat exchange equipment within the reactor pressure vessel was likely deemed necessary to achieve the natural circulation of the primary coolant in the PWR cycle.
A drawback of other PWRs that exist in the art is the fact that the reactor pressure vessels have penetrations at both the top portion of the reactor pressure vessel and at the bottom portion of the reactor pressure vessel. Still another drawback of existing PWRs is the fact that a substantial length of piping and a large number of joints are used carry the primary coolant from the reactor pressure vessel to the heat exchange equipment, thereby increasing the danger of failure due to a pipe break scenario.